ML20094C014
| ML20094C014 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 09/16/1977 |
| From: | Desiree Davis Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20094C015 | List: |
| References | |
| FOIA-95-262 TAC-01909, TAC-1909, NUDOCS 8707160815 | |
| Download: ML20094C014 (12) | |
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.fNITED STATES r-NUCLEAR REGULATORY COMMIS$10N I
.I WTASHINGTON. D. C. 20885
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4 NEERASKA PUBL:: POWER DISTRICT
- 0CKE~ NO. 50-298 4
COOPEk NUCLEAR STATION 1
AMENOMENT ~3 aC:L:TY OPERATING u: CENSE Amenament No. 38 License No. JPR 46
- .3
~ne Nuclear Degulat:ry : mmission 'the Conmissioni has found that:
A.
'he soplications #0r amenament by Nebraska Dublic Power District Ithe 'icenseet :atec Ceotemoer 10, 1975 ano January a, 1977, as sLoclementea cy 'etter catea April a,1977, comoly with the stancarcs and recu1rements of the Atomic Energy Act of 1954, as amenced ( the Act), ana tne Commission's rules and regula-tions set fortn d n 10 CFR Chapter I:
B.
~he facility will operate in confomity with the application, the provisions of the Act, ana the rules and regulations of tne Commission; 2.
nere is reasonaole assurance (i) that the activities authorized
~
,c Dy this amenament can De conductea without encangering the health ano safety of the public, ano (ii) that sucn activities will be conductea in comoliance with the Commission's regulations-2.
~he issuance of tnis amenament will not be inimical to the common defense ano security or to the health and safety of the public; ana E.
The issuance of this amendment is in accoraance with 10 CFR art il of :ne.c,mmission s regulations ano all applicaole recuirements have been satisfied.
2.
Accordingly, the license is amenced by a change to the Technical Specifications as inaicated in the attacnment to this license amenoment and paragraon 2.C(2) of Facility License No. DPR-46 is nereoy amenaea to reaa as follows:
2%. VM/B5 J
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r (2) Technical-Scecifications The "echnical Specifications containea-in Appendices A and B, as revisea nrougn Amendment No. 38, are hereby.
incorporatea in tne license.
The licensee snali operate the facility in accordance with the Technical Specifica-tions.
3.' This. license amenament 's effective as of the cate of its issuance.
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0 70R HE WCLEA~ REGULATORY COMMISSION
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Jon K. Davis, Acting Chief Operating Reactors Branch *2-Division of Operating Reactors Attacnment:
Changes to tne Tecnnical S peci fic a ti e.1s
'D Jate of !ssuance:
Sectemoer
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'977 1
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ATTACHMENT TO : ENSE AMENDMENT 'l0.
38 FACIL:TY OPEF.AT:NG LICENSE NO. DPR 46 20CKET '10. 50-298
- eplace existing : ages '62 ana 180 of the Aopendix A Technical 30eci#::a: ens ai:, :ne attacned evised Dages bearing the same
- hangec areas :n ne revisec ages are icentified by oumoers,
a marginal line.
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!MITING CONDITIONS FOR OPERATION SURVEII.I.ANCE REQUIRDfENTS t
. ". A (cent'd) 7.A (cont'd)'
repeated provided locally measured laakage reductions, achieved by re-pairs, reduced the containment's overall measured leakage rate suf-ficiently to meet the acceptance criteria.
f.
With the exception of main steam isolation valves and main steam line and feedwater line bellows, (see i
below) local leak rate tests-(LLRT's) shall be performed on the primary containment testa'le penetrations o
and isolatien valves at a pressure i
ef 58 psig during each reactor. shut-down for refueling but in no case
.at intervals greater than two years.
iolted double-gasket seals shall be-tested after each opening and during each reactor shutdewn for refueling but in no case at intervals greater than two years.
he main steam isolation valves (MSIV's) shall be tested a pres-sure of 29 psig.
If a total leak-age rate of 11.5 sci /hr for any one MSIV is exceeded, repair and retest shall be performed to correct the condition.
40 Main steam line and feedwater line
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i expansion bellows shall be tested at a pressure of 5 psig.
g.
Continuous Leak Rate Monitor When the primary containment is I
inerted, the containment shall be continuously monitored for gross l
1eakage by review of the inerting g
systc= makeup requirements.
This i
monitering system may be taken'out 4
of service for maintenance but shall be returned to service as soon as l
practicable, i
h.
Drywell Surfaces l
.he int erior surf aces or tne drywe12,,
I and torus shall be visually inspected,
each cperatin% cycle fer evidence of
- Exemption to Appendix J of lJ CFR 50.
Amendment No.138 e
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'*g JNITED STATEE NUCLEAR REGULATORY COMMISSION f-
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WASHINGTON. D. C. 20686 Yi SAFE ~v EVALUAT:UN SY nE ;FFICE OF NUCLEAR REACTOR DEGULATION SUPPORT:NG AMENDMENT NO. 23 0 FAC:LITY OPERATING LICENSE NO. DPR-46 AND EXEMPT:3NS ~; 10 CFR PART 50, APPENDIX J NEERASr.A PUBLIC POWER DISTRICT
- 0PER NUCLEAR STATION SOCKET NO. 50-298 0
- NT;000CT:0N s
o By letters Jatea Seotemoer ;0,1975 ana January 4, 1977, Nebraska Public Dewer ?i strict 'NPPD, the 'icense) ecuestec certain exemotions from the recuirements of Aopencix J to 10 CFR Part 50 for the Cooper Nuclear Station t:NS ).
~he requestea exemptions are:
1
~he ain Steam : solation /alves (MSIV's) wuld be testec I
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at 29 :sig (Pt) insteaa of the reauirea 58 osig (Pa).
2.
The personnel ' air ' ock coor would be tested at intervals
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'O no 1:nger tnan one year at 58 psig (Pa) ana at 3 psig af ter eacn opening curing the one year interval between
^g the 58 psig tests, j
i 3.
The volo between tne :eilows locatea in the main steam line anc feeawater line penetrations would be tested at 5 psig insteaa of tne requirea E8 psig (Pa).
4 The feedwater cnecK valves would be testec with water rather than air or nitrogen.
1 5.
The test interval for Type C valve leak testing would be extended until ne September 1977 refueling outage.
Adaitional infonnation concerning personnel air lock oor testing was
.provicea '- *.ne 'fcensee's loril A, 1977 'etter.
NPPD is currently evaluating our request for aacitional information on feedwater check
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valve testing.
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~.A 6. 7.A 3ASES_ (cont'd) trends.
Whenever a bolted double-gasketed penetration is broken and remade, tne space between the gaskets is pressurized to determine that the seals are performing properly.
- t is expected that the majority of the leakage from valves,penetrati:ns anc seals would be into the reactor building.
- However, it is possible that ;eakage into other parts of the facility could occur.
Such leakage paths that may affect significantly the consequences of accidents are to be =ini=1:ec.
Table 3. 7. 4 identifies certain isolation ualves that are tested by pressurizing the volume between the inboard and outboard isolation valves.
This results in conservative test results since the inboard valve, if a globe valve, will be tested such that the test pressure is tending to lif t the globe off its seat.
Additionally. the measurec leak race for such a test is conservatively assigned to botn of the valves equally and not divided between the two.
I The main stea anc feedwater testable penetratiens cuisist of a double layered
- etal bel;:ws.
The inbeard hi;h ;ressure side of the bellows is subjected to ir"wel: pressure.
~heretere tne bell:ws is tested in its entirety when the
- rewe;. :s tested.
~he tellews ;ayers are tested f:: the integrity ef both
. avers by pressuri:ing :he void between the ;ayers to 5 psig.
Any higher l
cressure ::uld :ause :er:anent defer =ation, damage and possible reptures of
- ne : ell:Vs.
The primary c:ntainment pre-operational test pressures are based upon the calculatec ort =ary containment pressure response in the event of a loss-of-coolant accident.
The peak drywell pressure would be about 58 psig wnich would rapidly reduce t 29 psig following the pipe break.
Following the pipe b reak, the suopression :hamber ;ressure rises to 27 psig, equalizes with drywell ;ressure and therefore rapidly decays with the drywell pressure decay.
.e desi:n :: es sure ci t he frywell and suppression cha:ber is 56 psig.
Based on the calculated containment pressure response discussed above, the primary
- entain ent re-coeratt:nal test eressure was chosen.
Also, based on the
'O prir.a ry centainment pressure response and the fact that the drywell and suppression chamber function as a unit, the primary containment will be s
tested a: a unit rather than the individual :caponents separately.
The design basis ;;ss-of-coolant accident was evcluated at the primary con-1eak race f 0.635'./ day at 58 psig.
tain=ent ax =um 211:wa:1e ace'
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Calculations cade c:. the NRC 5
.: wich leak rate and a standby gas treat-ment system filter ef ficiency :: 90* for halogens and assuming the fission eroduct release f ractt:.:s stated in KRC Regulaterv Guide 1.3, show that the max 1=um total whole body passing cloud dose is about 1.0 REM and the maximum t:tal thyroid dese is about 1: REM at 1100 meters f rc= the stack over an exposure duratien of two hetirs.
The resultant doses reported are the maximum that would be expectec in the unlikely event of a design basis loss-of-coolant accident. These doses are also based on the assu=ption of no holdup in the secondary contain=en-resulting in a direct release of fission products from the prt:ary
- ntainment tn:cugn the filters ano stack to the envirens.
Therefore. the specified pri=arv contain=ent leak rate and filter efficiency are canservative anc pr:vice cargin cetween expectec cif-site coses anc 10 CFR 100 cuidelines.
The water in the suppression chamber is used only for cooling in the event of an accident; i.e., it is not used for nor=al operation; therefore, a daily
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2 BACKGROUND Aooendix J to 10 CFR 50 was :u:lishea on February 14, 1973.
Since many operating nuclear plants nad either received an operating license or were in aavancea stages of design or construction at that time, some
- lants may not now be i" 'ul' :omoliance with the recuirements of this regulation.
Therefore, beginning in August 1975,' requests to establish the degree of compliance witn *he requirements of Appendix J were made of eacn licensee.
ollowing :ne initial responses to these requests, ae develooed ositions anicn ould provide assurance that the objective of the testing program were satisfied.
These NRC staff positions have since been apolier 'r, our eview of renor:s filed by NPPD anc the results
!*e -es::e
'" -.a #:ll: win; evaluation.
E. ALL *~ZN
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r e Itear 's: at'rr 'la i s e s of a :ena1x J reau1res :Sa ne containment isolation i
- ara;racr I:1.:.:
s
.aives :e locaii., 'ean testec ype C tes:; a: :ne peak calculateo con-
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31nnen: :res s ure, ;a.
Eacr main steam 'ine at CNS contains two containment isolation valves,
- ailec main stear sola:1:n valves : MSIVI in series, one inboaro valve anc cne cuttoarc valve with respec: to the reactor containment.
These valves are cesi:nea to :rovice a le+k tient seal in the main steam i
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'ines wnen :ne Calves are closeo and pressure is applied to the reactor
<essel side Cf the valve.
~herefore, if the MSIV's are shut in conjunc-
- 1on witn containment isolation, reactor vessei cressure or containment 4
- ressure, in :ne event Of a loss of coolant accident, on the valve i
so
..sc will hei; acnieve a tiont seal between :ne valve disc and seat.
s te :.rrent ;roceaure # r eak testing MSIV's at CNS requires pressuriz-
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'90 the main steam pipe volume Detween the inocard and outboard valves.
The procecure pressurizes the outboard valve in the direction for which
'; was designec 10 seal.
However, the procecure pressurizes the inboard valve in the reverse direction and, therefore, tends to open the valve Oy lifting the cisc off of its seat.
This' results in greater leakage
- nrougn the inboarc valve than would be experienced if the valve were pressurized in the proper direction.
The effect of reverse loading on
- ne inboard valve was considered when the original test pressure of 29 0519 was esta:lished'ana incorporated into *he CNS technical specifi-cations.
We have cete-rined that since the test pro:ecure used at CNS results in reverse I; acing of the 'nocarc MSIV ano theref:re in a greater measurea leak rate. testino MSIV's at 29 osic results in a conservative determina-
! * "' o f l e a k rate througn the vaives and is acceptaoie.
3-
?ersonnel Airlocxs Paragraons !I.'D.: ano ::1.3.2 of Appenoix s' require that reactor :entain.
ment airlocks :e lean tes ec at the-peak calculated accident pressure (Pa) at six On:n intervals.
Tur:ter, snould the airlocks be opened
' duriag sucn 'ntervals, :ne a:riccxs will be tes ec after each opening.
he c: ject;ve of :ne a1ricct 'eax esting recuirements are: (1) that the Jix rth tes:
~1 :revice an n:e; rated leakage rate for the entire asser:ly, inc?;;ing electri:3i anc mecnanical :enetrations, the airlock
- yi'n=er, ninge assemolies, oei:ec :onnections, anc ether potential
'eakage :atns; anc ',' :ta ne "adter each Openino" test would Orovide
'M a eans. f sssarirg :nat : e :: r seals nac no: :een damaged or seate:
1mpr::erly ::gring airlock ase.
The ai-loc,. :esign 'Or : e ::::er ita:1on 6ciu:es an inner and an outer docr. Octn
- *nten seat wita ::ntainment pressure.
Pressurizing the "1
airlock to Fa Iifts the. inner airlock coor off its seat which results in excess 1<e Camage nto :ne c:ntainment.
This concition does not reflect tae Oost-accident ::nci:i:- :# the airlock.
To leak test the airlock at Pa, a strongback rust be 'rstalled, insice the containment, on the inner airlock Ocor.
The stroa.goack crevents the inner door from lifting off its seat.
o ;oncuct :ne test, the airlock coors must be opened both tefore and a#:er ne tes
- nstall and remove the strongback.
Consecuently, NPPD has reauestec an exemotion to allow testing of the airlock at a recucec :ressure '3 :si:) wnicn would not require the use
'O of the strengcacn.
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1-We agree with N??D's Drecosed approach for the "after opening" airlock test.
Concutting the tests at ?a would necessitate breaking the coor seals 10 remove :ne strongeack, thus defeating the purpose of the tests.
Also, :ne 3 ;sig test :rovidea an acceptable test of the integrity of
- ne air lock door seals.
O
4 i-4 However, oefore we can conciace our review of this exemption request, we require an acceptance criterion for tne recucea pressure tests wnien correlates the leakage rate at 3 psig to the leakage rate which would be exoeriencea at 55 psig.
By letter dateo April 4, 1977, NPPD proviaea a correlation wnicn ecuates the 3 psig leak rate to the pro-auct of the 53 csig leak ate times the square root of the quantity three civicea Dy fi f ty-eignt.
-owever, NPPD providea no tecnnical aasis for inis eauation.
When NPPD orovides sucn a technical basis, we aill continue our evaluation of the 3 psig airlock test proceaure.
'PP) has also reauestec an exemetion to allow conoucting the airlock i
'ntegratec 'eak test at one year intervals ratner than at the six
,ontn rte-val s recuirec :y 'opencix J.
- nsuffi:ient justification was orovicea oy NPPD in suoport of a year test interval.
Accordingly, caseo on :ne lack of :ust*#icati:n, we #ind this crocosec exemotion unacceptacle.
s 7;
Main S team _ine ana Feecsater ' ine Bellows
?aragraon :::.5.2 of Appenaix, recuires that local leak test on con-tainment :enetrations (Type 9) ce performea at the ceak calculated
- ontainment oressure.
NPPD has requested an exemption from the Type
- 3 3 test-pressure for ne exoansion cellows in the main steam lines and feeawa ter 'ines.
The main steam ano feeawater testable penetrations i
- ensist of :oucle layerec metal cellows wnicn are currently locally
'eak testea ey pressur12ing the annulus Detween the couble layers to 9 :sig ratner tnan 33.
he cesign of the cellows aces not permit local testing at a nigner pressure.
The bellows are exposed to the c rywei l atmosonere ano are, therefore, testea as part of the contain-ment integratec 'eakage rate test.
In aadition, the bellows are a static system: there are no moving parts or active components.
Based on these considerations, we concluce that the proposed exemption for the cellows test pressure is acceptaole.
However, the NRC staff CI' 'eauire NPP3 to previce an acceptance criterion similar to that describec above for airlock testing to relate bellows leakage rates at 5 051g to the leakage rate which would be experiencea at 58 psig.
Feeawater Check Valves Paragraph :: 1.C.2 of Appendix. requires thut isolation valves be
'ocally leak testea wi th air or nitrogen.
3 Er :he :ncer ',uc: ear I:s ice :.e 'eeawater system was not originally designed :: :e r -1: 'ocal 'eak testing with air r r,itrogen.
- nerefore, eauestec an evemot on 'or the feeawater eneck valves which i
3re :urrently 'eak teste:.<i.- aater 'nsteaa cf air or nitrogen as recuirea :y Accenaix J.
This paragraor, :f icoenc's ? -ecaired a simulation of the conoition of the system following 5 ;Csislatea loss-of-coolant accident (LOCA) wnere :ne leakage carriers e.g., valves, gaskets, and seals) may be (xoosea : tne c:ntaintren. atmes:nere.
~here are a numoer of liquid-
'illec systers. owever. na: are :esignec :o remain intact following a LOCA.
~'ese 'iauia 'ea systems include :ne emergency core cool-d ng s ste anc :ne c:ntain ent neat removal systems.
For those 5 stems :nat are :esi:nec n engineerea safety #eature criteria and f;r anien :nere s asserarte ha: they will remain filled with liquic
- i 0winc a L;:A :ne :utc 'eakace rates snould be cistinguished 7
- :- ::U.a' : en: a; s:cere
+a.. age rates.
~hese systems can ce nyaresta icall, :estea u :emonstrate that :ne #1uid inventory is
- uf#i: en: u.a.ntai n a.sa'er seal during ano following the postulated ac 1:ent.
. ii:Lia 'easage 'imit can be assignec #cr these systems.
Th': :-1:erton is similar - concept to a valve seal-water system
'O criter :n anc wil' :rovice eouivaient isolation rotection.
For this type of testing, -aalological analyses should :e performed to demon-strate : a: :ne 'icule 'eaxage limits do not result in significant
- ses to na:. e u tal 2::icen
- ose would not :e greater than the
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- :R :te: ':: guide 1ines.
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- - :s Ie::emcer ':.
'9"'E :; m10:si, NPPD reauested an exemption from
- ,e recu1remeru :f Appen:1x to permit locai 'eak rate testing of s
- e 'eecwate- : e:L.'ai.es ; sing aater as a test mecium rather than Sir :r -1:-: en.
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ae r ave :ete r-inea
.a t : e :rocesea hydrostati: esting would be 2cceptatie 1'
can te E own 19et the valves will indeed be filled witn water our'ng ana aner a loss-of-coolant.LOCA) accident and that ne 14 aut : leatage u i'
- -esult n aaditional radiological deses i
sucn :nat :ne a:al acc1:ent dose would be greater than the 10 CFR
?ar: 100 guicelines.
By letter cateo Feoruary 17, 1977, we requestea that NPPD demonstrate
- nat :ne #esawater, ana -PC: anc RCIC-to-feeawater, check valves would remain 'a',1 :.f aater following a postulated LOCA ana that the fission products entrainea in liauid leakace will not result in total radio-logical uses exceecing '3 CFR Fart 100 guicelines.
Alternatively, we
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askee NP:: u eitner :eveioo a ::rreistion to an ecuivalent air
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leakage or moc1fy :ne systems :: ;ermi: :eax :esting with air er nitrogen.
When the licensee Or: vices :ne -ecuestec d nformation, we will c:ntinue our evaluation of this exeme:icn recuest.
Type C Test ' e.t e r va l Paragraon :: 1.2.2 :f ::ercir. recuires. a: Type C tests be ;erformed curing eacn reac r snu ::wn #:r refuelin; :ut in no case at intervals gr. eater than wo years.
~he e:ari:si ::ec #i:a:i:ns #:r ~;0cer Nuclear Sta:i:n
.N5' 5:ecify that ~ype C tests snall :e :er# rmee eacr ::erating cycle Out in no a
case at intereals ;reater var two.s ea rs.
2 As statec :n NF:0's Janua ry 4, :977 letter, curing the re'ueling of CNS wnien ::curre: :etoeen ie *.e :er 15 anc Ncvemcer iC, 1976, Type C tes:-
ing was not :er#:r ec :eca,se "ac been :erformec during a naintenance outage in Oct::er 1975 anc aas scneoulea :: be performea during :ne next refueling outage in Seo em:er 1977 wnicn :s within :ne 24 montn period requirec Oy :::n : 0 ;?; 53 :::encix J anc :ne CNS Tecnnical Specifications.
In discussions suosecaent to :ne startuo :# CNS following the October 1976 refueling outage. -e ncicatec tna: ND0's interpretation of tne 3
Type C test ##ecuenc;. aas ncorrect; anc :ne Type C testing should have d
been per#0rmec curir: :na: -e#uelin: outate.
Therefore. in a letter catec January 4, 1977, NPPO reauestec an exemotion to permit :ne perfor-mance of "ype C :ests curing :ne refueling Outage seneauled for Septem:er I;77.
1 The frecuency # r Type C tests soecifiec 'n Accencix J was selected to coincice witn *.ne refueling outage wnicn is normally not more than two years after *.ne first efueling.
This is because a shutcown and cooldown solely #:r *.nese tests would esult in an ;nnecessary plant thermal cycle.
Sucn thermal :ycles are limitec by cesian to minimi:e the effects of thermai ana mecran1:si stresses on plant systems.
Therefore, it is desirable :: ::nc : "ese tests curin 50me other scheculed shutdown and cooldown event, sucn as refueling,'bu: in no case at intervals greater : nan two years, Because approval of this exemotion would not result in an unnecessary thermal cycle or exceecing the maximum s:ecifiea test interval and because approval would result *n placing CNS back en the inspection scnedule specifiec in Appencix J, we concluce tnat :ne proposed one-time exemo-
- ion is acce::aole.
- wever, :o :revent #;ture misinterpretation, tecnnical specification 4.7. A.f will be enanged to bring it into verbatim agreement with caragraph III,D.3 Of Appendix J.
E nvironmental T ons1 aeration a'e. nave aeterminea that :ne amenament coes not. authorize a cnange in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
9aving mace this cetermination, ae have further concluded that the amenament-involves an acti:n anics is insignificant from the stand-point of environnental imoac: anc pursuant to 13 CFR 551.5(o)(4) that an environmental 'mcact statement, Or negative declaration ano environemntai 'moact aopraisal neea not ce :repared in connec-tion ai tn :ne isssance of :nis amenament.
- onclusion
.o I
3asec on tne #:regoing, e ave :eterminea :nat, pursuant to 10 CFR Iection 50.12, specific exempti:ns for *SIV testing, steam anc feed-i' water line cellows testing, 2na Type C test interval, as aiscussed
)
aDove, can ce grantec ai:nout encangering life or property, or the common cefense ana security, ano are otnerwise in the public interest.
l We nave also conclucea, : asea on :ne considerations discussea above, that:
- 1) Decause tne amenenent :ces not i.volve a significant fncrease in the :rocability or :ensecuences of accidents previously
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consiaerea ana cces not involve a significant cecrease in a safety margin, :ne amenament aces not involve a significant hazards 1
wa
- ensiceration..2) nere 's reasonable assurance that the health ina safety ;f :ne :uolic.111 not de encangerea cy operation in the
,3 prooosea manner, ano (3) such activities will be conoucted in com-1 pliance with tne Commission's regulations ano the issuance of this
)
amenonent will not ce inimical to the common cefense and security or to the neal th anc Safety of tne ouDlic.
a te : Ee::emoer
'5,
'.977 I
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-